1. Field of the Invention
This invention relates to an improved compact model steam generator which has the ability to accurately simulate a variety of separate thermo-hydraulic conditions within a full-scale steam generator in order to separately monitor the corrosion effects of these conditions on the heat exchange tubes within the full-scale steam generator.
2. Description of the Prior Art
Model steam generators for monitoring the amount of corrosion degradation occurring within the heat exchange tubes of nuclear steam generators are known in the prior art. Generally speaking, such model generators operate by subjecting an array of sample heat exchange tubes to the same set of heat, pressure and chemical conditions which surround the heat exchange tubes in such nuclear steam generators. If these conditions are accurately simulated, the amount and nature of corrosion which occurs in the sample tubes of the model steam generator will provide an accurate indication of the tube corrosion present in the nuclear steam generator being monitored. Such model steam generators are a particularly useful form of corrosion monitor because they obviate (or at least reduce) the need for shutting down a nuclear plant and sending technicians into the radioactive interiors of the generators.
However, such model steam generators are useful only insofar as they are capable of accurately simulating at least one set of the heat, pressure and chemical conditions which actually exist within a selected portion of the nuclear steam generator. Any material departures from these conditions will adversely affect the ability of the model steam generator to accurately monitor the amount of corrosion accumulating around the heat exchange tubes and support plates within the full scale steam generator. In order to understand the difficulties in building a practical model steam generator which provides an accurate monitor for such tube corrosion, one must first understand how nuclear steam generators are generally constructed, and what chemical, thermal and hydraulic conditions are responsible for the tube corrosion.
Nuclear steam generators are comprised of three principal parts, including a primary side, a secondary side, and a tubesheet in which the inlet and outlet ends of a plurality of U-shaped tubes are mounted. The tubesheet and U-shaped tubes define a pressure boundary between the primary and secondary sides. The primary side of the generator defines a first hydraulic flowpath through the inlets, outlets and interior surfaces of the U-shaped tubes, and includes a divider sheet which hydraulically isolates the inlet ends of the U-shaped tubes from the outlet ends. The secondary side includes a feedwater inlet, and defines a second hydraulic flowpath around the outside surfaces of these tubes. Hot radioactive water flowing out of the nuclear reactor core is admitted into the section of the primary side containing all of the inlet ends at the U-shaped tubes. This hot water flows into these inlets, up through the tubesheet, and circulates around the U-shaped tubes which extend within the secondary side of the steam generator. The heated water transfers its heat through the walls of the U-shaped tubes to non-radioactive feedwater and recirculating water flowing through the secondary side of the generator, thereby converting it to non-radioactive steam. After the nuclear-heated water circulates completely around the U-shaped tubes, it flows back through the tubesheet, through the outlets of the U-shaped tubes, and into the outlet section of the primary side, where it is recirculated back into the core of the nuclear reactor. The inlet ends of the U-shaped tubes are known as the "hot legs" and the outlet ends of these tubes are known as the "cold legs". An illustration of this general arrangement is present in FIG. 1 of U.S. patent application Ser. No. 567,328, filed Dec. 30, 1983 and assigned to Westinghouse Electric Corporation, the entire specification of which is hereby expressly incorporated herein by reference.
Over a period of time, the heat exchange tubes of such nuclear steam generators can suffer a number of different types of corrosion degradation, including denting, stress corrosion cracking, intergranular attack, and pitting. In situ examination of the tubes within these generators has revealed that most of this corrosion degradation occurs in what are known as the crevice regions of the generator. Such crevice regions include the annular space between the heat exchange tubes and the tubesheet, as well as the annular clearance between these tubes and the various support plates in the secondary side which are used to uniformly space and align these tubes. It is believed that the corrosion which occurs in these crevice regions is caused from the corrosive chemicals in the sludge which accumulates in these regions. Deposits of sludge tend to collect in these crevices from the effects of gravity. Additionally, the relatively poor hydraulic circulation of the water in these regions tends both the create and to maintain the sludge in these crevices, as well as to create localized areas of elevated temperature (or "hot spots") in the tubes adjacent the sludge. These "hot spots" create local concentrations of corrosive impurities and act as a powerful catalyst in causing the exterior surface of the heat exchange tubes to react with the corrosive chemicals and the sludge. The resulting corrosion products tend to fill the crevices even more, thereby exacerbating the corrosion-producing conditions. While most nuclear steam generators can be sludge-lanced to periodically sweep the sludge out of the generator vessel, the sludges in the crevice regions are not easily swept away by the hydraulic currents produced by such systems. Despite the fact that the heat exchange tubes of such nuclear generators are typically formed from corrosion-resistant Inconel.RTM. 600 alloy, the combination of the localized regions of heat and corrosive sludges can ultimately cause the heat exchange tubes to crack, and leak radioactive water from the primary side into the non-radioactive water in the secondary side of the generator. However, such dangerous leakage need not occur if the heat exchange tubes are subjected to remedial action (such as plugging or sleeving) before corrosion causes cracks in the walls.
Model steam generators were developed in order to accurately monitor the amount of corrosion degradation occurring in the heat exchange tubes of a particular nuclear steam generator so that corrective actions may be taken before any of the tube walls crack. Such model steam generators have been found to be a particularly accurate way of ascertaining the amount of corrosion degradation occurring in the heat exchange tubes of a nuclear steam generator, because the particular amount of corrosion which the feedwater chemistry and thermohydraulics in a particular region of a given generator will induce in a particular set of tubes is virtually impossible to predict by purely theoretical models.
Unfortunately, prior art model steam generators are not without significant shortcomings. For example, none of these prior art model steam generators includes more than one primary system. Accordingly, if the plant operator wishes to simultaneously monitor the corrosion effects on heat exchange tubes in different regions of the full-scale generator, he must purchase and install two separate model steam generators if he is to obtain the information he desires. Another shortcoming associated with such prior art model steam generators is their relatively large size and weight, which makes them unwieldy not only with respect to installation, but to disassembly and reassembly after the completion of each monitoring test. One type of model steam generator which solves much of the size and weight problems by means of a simple and relatively inexpensive design is described and claimed in U.S. application Ser. Nos. 636,437, 636,438, 636,449 and 636,450, all of which were filed on July 31, 1984 and assigned to the Westinghouse Electric Corporation. However, this particular design of model steam generator still does not have the capacity to simultaneously simulate two or more sets of thermo-hydraulic conditions which may exist within the full-scale steam generator being monitored. Accordingly, a power plant operator wishing to simultaneously monitor the tube corrosion in both "hot leg" and "cold leg" portions of the full-scale generator would have to purchase, install and operate two of these types of model steam generators. Clearly, there is a need for a model steam generator which is capable of simultaneously simulating two or more sets of thermo-hydraulic conditions in order that the amount of tube corrosion occurring in different regions within the full-scale generator may be monitored. Ideally, such a model generator should also be relatively compact in size so that it could easily fit into areas of constricted space, and lightweight so that it could be easily installed on the feedwater system of the full-scale generator being monitored. Finally, it would be desirable if it were also easily dissembled and reassembled in order that the operators of the model generator might conduct their corrosion-monitoring tests with a minimum of awkward and time-consuming handling.